This invention relates to nuclear reactor power monitoring systems for monitoring the power level of a reactor and preventing an excessive rise thereof, and more particularly it is concerned with a system for controlling the core coolant flow rate of a nuclear reactor particularly designed for preventing an excessive rise in power level attributable to a transient increase in the core coolant flow rate.
Generally, control of the power level of a boiling-water nuclear reactor (BWR) is effected not only by practicing selective operation of a plurality of neutron absorbing control rods which are movable into and out of the reactor core, but also by effecting adjustment of the core coolant flow rate in combination with the selective operation of the control rods.
The principle of power control of BWR relying on the adjustments of the core coolant flow rate is that the relation between variations in the manner of production of voids in the core caused by changes in the core coolant flow rate and the negative reactivity coefficient of voids can be utilized for controlling reactor power. More specifically, in increasing the nuclear reactor power, the core coolant flow rate is increased to increase reactivity as the volume of voids in the core is transiently reduced, thereby increasing reactor power. This increase in reactor power continues until balance is achieved between the amount of the increase in reactivity in the transition period and the amount of voids newly produced in the core. Thus a new reactor power level is attained.
The core coolant flow is created by leading part of the coolant out of the pressure vessel and driving recirculation pumps which forces the coolant to flow into jet pumps within the pressure vessel. The core coolant flow rate is controlled by a flow control system associated with the recirculation pumps. The core coolant flow does not show the same rate as the recirculation flow. However, there is a uniform functional relation between these two values, so that the core flow rate can be indirectly controlled by controlling the recirculation flow rate. Control of the recirculation flow rate can be effected either by adjusting the opening of flow control valves at the discharge side of the recirculation pumps or by controlling the number of revolutions of the recirculation pumps.
Operation of BWR is practiced in accordance with a predetermined operation plan by the aforesaid selective operation of the control rods and the control of the core coolant flow rate. The operation plan is determined in such a manner that reactor power can be maximized without impairing the integrity of fuels by an excessive rise in power. Core characteristics parameters regarding the integrity of fuels include a maximum linear heat generating rate which is a monitor index for preventing mechanical breakdown of the fuel cladding due primarily to thermal deformation and a minimum critical power ratio (MCPR) which is a monitor index for preventing thermal breakdown of the fuel cladding caused by transition boiling or film boiling of the coolant, both being controlled in such a manner that they do not exceed the respective critical levels that would cause breakdown of the fuel cladding.
During operation of BWR, reactor power may deviate from the aforesaid operation plan for some reason and give rise to an excessive rise in power. To avoid this phenomenon, a core monitoring system is provided. The conventional power monitoring system includes an average power range monitor (APRM), a thermal power monitor (TPM) and a rod block monitor (RBM). These monitors all operate as follows on the basis of signals from a neutron monitoring system including local power range monitors (LPRMs). APRM monitors the power level of the nuclear reactor which is obtained by averaging the local power signals from all LPRMs, and prevents withdrawing of control rods when the power level attained by withdrawing of control rods has reached a predetermined rod block threshold and scrams the nuclear reactor when the power level has reached a predetermined scram threshold as a result of withdrawing of control rods, as increase in the core coolant flow rate and the core pressure, etc. TPM monitors thermal power transferred through fuel cladding surfaces, the thermal power being converted from local power signals from LPRMs through a time delay circuit, and scrams the nuclear reactor when the thermal power has reached a predetermined scram threshold. RBM monitors changes in the readings of LPRMs adjacent to the control rod to be withdrawn and prevents withdrawing of the control rods when the readings reaches a predetermined rod block threshold. APRM and RBM are disclosed, for example, in U.S. Pat. No. 3,565,760 granted to G. R. Parkos et al on Feb. 23, 1971.
In the core monitoring system described hereinabove, withdrawing of control rods is blocked either by APRM or RBM when the power level of the core or the local power level thereof attained by withdrawing of control rods reaches the respective threshold. Thus APRM and RBM suppress an excessive rise in reactor power caused by withdrawing of control rods but do not provide means for suppressing an excessive rise in reactor power caused by a change in the core coolant flow rate. The suppression of the latter excessive power rise is effected by APRM and TPM in the existing circumstances, which scram the nuclear reactor when the power level or thermal power level of the core reaches the corresponding one of their scram thresholds. APRM and TPM also scram the reactor when the reactor power level reaches the corresponding one of their scram thresholds as a result of withdrawing of control rods.
The nuclear reactor continues its operation when withdrawing of control rods is blocked by APRM or RBM, so that the blocked control rods can be actuated again if reactor power is reduced by inserting other control rods or reducing the core coolant flow rate. However, in the event of the core flow rate being increased and the reactor power level being raised by an operator turning some wrong valves or some equipment misoperating, for example, the power level only rises until the scram threshold of APRM or TPM is reached as aforesaid and the reactor is scrammed. As a result, the number of times the reactor is scrammed increases more than is necessary and interfers with the operation of the reactor. Also, when the reactor power level rises as aforesaid, the aforesaid core characteristics parameters such as the maximum linear heat generating rate and minimum critical power ratio may exceed the critical level that may cause breakdown of the fuel cladding. To avoid this accident, it has hitherto been necessary to set the power level of normal operation of a nuclear reactor at a relatively low level to provide reserves for this accident.
Japanese Patent Publication No. 21518/79 published on July 31, 1969 for Tokyo Shibaura Electric Company, Ltd. discloses a control system in which (1) in a slow power increase mode a recirculation coolant flow rate increase rate m or a core coolant flow rate increase rate c is limited to a level lower than the level obtained by calculation performed as predetermined, and (2) in a normal operating mode, a recirculation coolant flow rate threshold M and a core coolant flow rate threshold C are reset only when power density is found to be higher in level than the value obtained at the preceding calculation as calculation thereof is carried out at certain time intervals, to thereby avoid an increase in flow rate above the reset threshold.
In this control system, when the core flow rate falls and power is reduced after the thresholds M and C are reset at a high power level P.sub.H following a slow rise in power, the reset thresholds M and C are kept at a high level. If a rise in flow rate is caused, for example, by the failure of a recirculation flow controller after the power is kept at a low level for some time, the flow rate continues to rise until it reaches the level of M or C. The power level attained at this time would be higher than the power level P.sub.H by an amount corresponding to a reduction in Xenon (neutron absorber) in the core occurring during the time of low level operation of the reactor. Thus, the process for setting the thresholds shown in Japanese Patent Publication No. 21518/79 is such that a flow rate threshold is set at the power level P.sub.H and the threshold, once set, may be kept constant irrespective of power until the power level exceeds P.sub.H again and a fresh threshold is set following recalculation of power density by the computer. Therefore, even if the flow rate rises to the threshold again in the interval due to a reduction in power and a variation in the amount of Xenon, there is no assurance that power will be lower than the threshold P.sub.H. In fact, there is no express mention of the amount of Xenon in formulas (1)-(12) described in Japanese Patent Publication No. 21518/79.